The invention relates to a method and system for monitoring at least one operating parameter of the core of a nuclear reactor and in particular a pressurized water reactor.
Nuclear reactors such as pressurized water reactors have a core consisting of fuel assemblies which are generally straight, of prismatic shape and juxtaposed with their longitudinal axis vertical, i.e. in the direction of the height of the core.
It is essential to ensure at all times that the reactor is operating perfectly and in accordance with general safety conditions set by regulations and standards.
In particular, it is necessary to determine if the production and spatial distribution of the flow of neutrons and the spatial distribution of the power generated in the core conform to conditions corresponding to normal and satisfactory operation of the core.
This requires calculation of operating parameters of the core of the nuclear reactor, such as the spatial distribution of power in the core, the form factors of the neutron flux or the critical heating ratio. These parameters are determined by measuring the neutron flux in the core to determine the distribution of the neutron flux throughout the core in three dimensions.
The parameters characteristic of the state of the core in normal operation which are derived from the neutron flux measurements must not at any time be outside ranges determined during the process of designing the nuclear reactor.
It is necessary to trip an alarm and to implement various measures concerning the control of the nuclear reactor if any parameter characteristic of the operation of the core is found to exceed a limiting value.
For effective monitoring of the operation of the core of the nuclear reactor it is necessary to determine the operating parameters of the core and therefore the distribution of the neutron flux in the core in as short a time period as possible.
The neutron flux measurements in the core needed for continuous monitoring of the nuclear reactor in operation are generally provided by chambers outside the containment vessel of the reactor which are generally referred to as xe2x80x9cexcorexe2x80x9d chambers.
These chambers include multiple measuring stages (for example six stages) distributed in the direction of the height of the core and are generally used to perform measurements in four areas at the periphery of the core of the nuclear reactor which are symmetrical about two axial planes of symmetry of the core at an angle of 90xc2x0 to each other.
The staged excore detector chambers provide flux measurements at various heights in the core and in these four circumferential areas around the core. However, these external systems provide only approximate values of the neutron flux within the core and an approximate representation of the neutron flux distribution. Accordingly, the monitoring parameters are obtained in a relatively imprecise manner and, for safety reasons, greater margins must be provided for critical values of these parameters that must not be reached or exceeded.
To obtain a more exact representation of the neutron flux distribution within the core, additional neutron flux measurements are conducted within the core, at regular but relatively long time intervals, for example in the order of one month, using very small measuring probes, referred to as xe2x80x9cincorexe2x80x9d probes, which generally take the form of fission chambers. Each incore probe is fixed to the end of a flexible cable referred to as a teleflex cable for moving it inside a measuring channel of the instrumentation of the nuclear reactor. Each measuring channel opens at one end into an instrumentation area in the bottom part of the reactor building. The fission probes are moved inside the measuring channels from the instrumentation area. Each measuring channel includes a fuel assembly instrumentation tube inside the core of the nuclear reactor and a glove finger inside the instrumentation tube in which the fission probe moves. The neutron flux is measured in a set of fuel assemblies distributed throughout the section of the core.
In the case of a core with 177 fuel assemblies, 56 measuring channels are generally used, for example. Similarly, 58 measuring channels are used for a core with 193 fuel assemblies, 50 measuring channels for a core with 157 fuel assemblies and 60 measuring channels for a core with 205 fuel assemblies. The neutron flux measurements are carried out as the incore probes are moved at slow speed over the full height of the core. Many neutron flux measurement points closely spaced along the height of the core are obtained in this way. An image which is sufficiently representative of the neutron flux is obtained in the form of a flux map, given the distribution of the instrumented fuel assemblies within the core and the symmetry of the core. However, the incore probes consisting of the fission chambers cannot be used for very long periods inside the core of the nuclear reactor. An accurate flux map of the core is determined only from time to time and therefore cannot be used for continuous monitoring of the operation of the core of the nuclear reactor.
Also known in the art are neutron flux measuring probes which can remain in the core of a nuclear reactor throughout the operation of the nuclear reactor. These neutron flux measuring probes, which can take the form of xe2x80x9ccollectronsxe2x80x9d, are generally assembled in the form of measuring rods in a vertically aligned arrangement with a constant spacing between two successive probes, to constitute flux measuring detectors distributed throughout the height of the core of the nuclear reactor. Each rod is inserted into a glove finger normally used for measurements by a mobile probe and itself inserted into the instrumentation tube of a fuel assembly. Each of the flux measurement detectors or measurement rods, whose length is almost equal to the height of the core, can include eight measuring probes in the form of collectrons, for example.
In the case of a nuclear reactor core with 177 fuel assemblies, it has been proposed to place 52 measuring rods or detectors in 52 instrumented assemblies of the core of the nuclear reactor distributed throughout the section of the core.
An instrumentation system of the above kind, which has 8xc3x9752 measuring points distributed throughout the core, is able to provide an accurate three-dimensional image of the flux distribution in the core of a nuclear reactor.
However, given the response times necessary for monitoring the core of the nuclear reactor, processing neutron measurements produced by instrumentation remaining permanently in the core while the nuclear reactor is operating takes a long time, which is hardly practicable for monitoring operating parameters of the core.
Thus there is no accurate and fast method of obtaining the spatial distribution of the flux and of obtaining parameters for monitoring the core of the reactor from the flux distribution.
Nuclear power stations generally have a plurality of units or xe2x80x9ctranchesxe2x80x9d each consisting of a nuclear reactor in a reactor building and conventional electrical power generation part. In this case, core monitoring concerns the nuclear reactor of each tranche.
The object of the invention is therefore to propose a method of monitoring an operating parameter of the core of a nuclear reactor of a tranche of a nuclear power station consisting of a plurality of fuel assemblies juxtaposed along the height of the core using a set of neutron flux measuring detectors introduced into at least some of the fuel assemblies of the core and each including a plurality of fixed neutron flux measuring probes distributed along the height of the core, the method enabling fast and accurate determination of the flux distribution in the core, fast and accurate determination from the flux distribution of an operating parameter of the core, and substantially real time monitoring.
To this end, while the nuclear reactor is operating, at particular time intervals:
the spatial distribution of the neutron flux is measured using a subset of neutron flux detectors selected from the set of detectors introduced into the fuel assemblies of the core to obtain measurement signals, the number n of detectors of the subset being at most equal to 15% of the number of fuel assemblies in the core,
a code for computing the neutron flux is used and account is taken of the measurements supplied by the subset of neutron flux detectors to obtain the instantaneous distribution of the, neutron flux in the core in the form of a set of values of the neutron flux at points distributed through the core,
at least one operating parameter of the core is calculated from the instantaneous neutron flux distribution, and
an alarm is triggered if at least one parameter is outside a fixed range.
In a first embodiment of the invention, to obtain the instantaneous neutron flux distribution throughout the core:
the spatial distribution of the flux in the core is calculated instantaneously on the site of the reactor from parameters from equipment of the tranche, using the neutron flux calculation code and in the form of a set of neutron flux values at points distributed through the core including a first subset of instrumented positions at which the n neutron flux measuring probes of the subset of detectors are situated and a second subset of non-instrumented positions,
the differences between the measured flux values and corresponding values calculated from the parameters from the equipment of the tranche are computed for each instrumented position,
corresponding differences are calculated for each non-instrumented position by extrapolation from the differences relating to the instrumented positions, and
an algebraic value is added to the spatial flux distribution values obtained from the parameters from the equipment of the tranche to reduce the calculated difference at each of the points distributed in the core .
In a second embodiment of the invention, to obtain the instantaneous neutron flux distribution throughout the core:
the spatial distribution of the neutron flux inside the core is calculated instantaneously a first time, on the site of the reactor, from parameters from the equipment of the tranche, using the neutron flux calculation code and in the form of a set of neutron flux values at points distributed throughout the core including a set of instrumented positions at which the neutron flux measuring probes of the subset of detectors are situated,
the differences between the measured values of the neutron flux and corresponding values calculated from the parameters from the equipment of the tranche are calculated for each instrumented position,
the calculated differences are used to apply a correction to definition parameters of the neutron flux calculation code, and
the instantaneous distribution of the neutron flux inside the core is calculated instantaneously a second time, on the site of the nuclear reactor, from the parameters from the equipment of the tranche and using the neutron flux calculation code including the corrected definition parameters.
In order to explain the invention, one embodiment of the method according to the invention will now be described by way of example and with reference to the accompanying drawings, in the context of a pressurized water nuclear reactor and its monitoring system.